• Transactions of the Korean Society of Pressure Vessels and Piping
• /
• v.9 no.1
• /
• pp.20-24
• /
• 2013

Reactor Vessel Internals(RVIs), which are installed within the reactor pressure vessel and support the fuel assembly, take responsibility for safety of reactor core. In operating Nuclear Power Plants(NPPs), the RVIs have been exposed to severe conditions such as neutron irradiation, high temperature, high pressure, and high velocity of coolant flow and have degraded by materials aging with long-term operation. Therefore, the effective aging management plan and the appropriate regulatory requirements are necessary to maintain the integrity of RVIs. The purpose of this paper is to provide a review guide for Periodic Safety Review(PSR) of RVIs in presurized water reactor. The review guide is developed based on the revised review guides and reports established from IAEA and USNRC, and the analysis results of design characteristics, aging mechanisms, and operating experiences of RVIs in domestic and international NPPs. Consequently, the developed review guide for PSR of RVIs is expected to contribute an overall strategy and standard for the PSR of RVIs.

• Proceedings of the Korean Nuclear Society Conference
• /
• 2001.05a
• /
• pp.461.1-461
• /
• 2001

• Proceedings of the Korean Nuclear Society Conference
• /
• 2005.10a
• /
• pp.65-66
• /
• 2005

• Nuclear Engineering and Technology
• /
• v.28 no.6
• /
• pp.576-582
• /
• 1996

The integrity of reactor vessel support system of the Korean Standard Nuclear Power Plant (KSNPP) is investigated for a postulated reactor vessel closure head drop event. The closure head is disassembled from the reactor vessel during refueling process or general inspection of reactor vessel and internal structures, and carried to proposed location by the head lift rig. A postulated closure head drop event could be anticipated during closure head handling process. The drop event may cause an impact load on the reactor vessel and supporting system. The integrity of the supporting system is directly relevant to that of reactor vessel and reactor internals including fuels. Results derived by elastic impact analysis, linear and non-linear buckling analysis and elasto-plastic stress analysis of the supporting system implied that the integrity of the reactor vessel supporting system is intact for a postulated reactor vessel closure head drop event.

• Journal of KSNVE
• /
• v.7 no.1
• /
• pp.33-41
• /
• 1997

To cover a range of possible site conditions where the Korean standard nuclear power plant may be constructed, a range of generic site conditions is selected for geologic and seismologic evaluation. To envelop other Asian countries as well as the Korean peninsula, there is an attempt to increase the seismic level to 0.3g ground motions for the safe shutdown earthquake. The dynamic analyses of the reactor vessel internals and fuel assemblies are performed for the increased motions and the effect of seismic level on the response is investigated. Also the nonlinear response characteristics are discussed by comparing the loads between operating basis earthquake and safe shutdown earthquake excitations. The design adequacy of the reactor vessel internals and fuel assemblies for the increased seismic level is addressed.

• Transactions of the Korean Society for Noise and Vibration Engineering
• /
• v.21 no.8
• /
• pp.708-713
• /
• 2011

U.S. nuclear regulatory commission(NRC) regulatory guide(RG) 1.20 requires a comprehensive vibration assessment program(CVAP) for use in verifying the structural integrity of reactor vessel internals(RVI) for flow-induced vibrations prior to commercial operation. The CVAP program consist of vibration and fatigue analysis, a vibration measurement program, an inspection program, and a correlation of their results. One of the main purposes of the analysis program is to select measurement locations, however measurement locations can not be determined by only analysis results, therefore we developed selection criteria of measurement locations for advanced power reactor 1400(APR1400) RVI CVAP, It will be used to select measurement locations and instrument types for APR1400 RVI CVAP.

• Nuclear Engineering and Technology
• /
• v.12 no.4
• /
• pp.286-292
• /
• 1980

Similitude requirement for model testing of flow induced vibration of reactor internals are investigated. In depth discussions on the Reynolds number effects are made. For valid model tests of fuel assemblies vibrating in its fundamental natural frequency, reduced frequency (fD/U), and dam ping parameter( $m_{c}$$\delta$$_{c}$ $D_{\rho}$$^2$) are two most important parameters.ers.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
• /
• 2014.10a
• /
• pp.102-105
• /
• 2014

In accordance with U.S. Nuclear Regulatory Commission regulatory guide(NRC RG) 1.20(Rev.3), we are writing a comprehensive vibration assessment program(CVAP) design control document(DCD) and a technical report for U.S. NRC design certification(DC) of an Advanced Power Reactor 1400(APR1400) nuclear power plant(NPP). CVAP of an APR1400 NPP for U.S. NRC DC is classified as a non-prototype category 1 type. Therefore, CVAP DCD of reactor vessel internals(RVI) and steam generator internals(SGI) consist of analysis and full inspection program. However, piping system of primary and secondary system will be described as measurement program.

• Nuclear Engineering and Technology
• /
• v.55 no.1
• /
• pp.80-99
• /
• 2023

This paper presents the results of investigating the change in welding residual stresses of the core shroud, which is one of subcomponents in reactor vessel internals, performing finite element analysis. First, the welding residual stresses of the core shroud were calculated by applying the heat conduction based lumped pass technique and finite element elastic-plastic stress analysis. Second, the temperature distribution of the core shroud during the normal operation was calculated by performing finite element temperature analysis considering gamma heating. Third, through the finite element viscoelastic-plastic stress analysis using the calculated temperature distribution and setting the calculated residual stresses as the initial stress state, the variation of the welding residual stresses was derived according to repeating the normal operation. In the viscoelastic-plastic stress analysis, the effects of neutron irradiation on mechanical properties during the cyclic normal operations were considered by using the previously developed user subroutines for the irradiation agings such as irradiation hardening/embrittlement, irradiation-induced creep, and void swelling. Finally, the effect of neutron irradiation on the welding residual stresses was analysed for each irradiation aging. As a result, it is found that as the normal operation is repeated, the welding residual stresses decrease and show insignificant magnitudes after the 10^th refueling cycle. In addition, the irradiation-induced creep/void swelling has significant mitigation effect on the residual stresses whereas the irradiation hardening/embrittlement has no effect on those.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
• /
• 2012.04a
• /
• pp.734-738
• /
• 2012

A comprehensive vibration assessment program for the Advanced Power Reactor 1400 reactor vessel internals is established in accordance with the United States Nuclear Regulatory Commission Regulatory Guide 1.20 Revision 3. This paper is related to instruments and measurement locations based on the vibration and stress response analysis results at the inner barrel assembly top plate in the reactor. The analysis results of the inner barrel assembly top plate in the reactor show that the deterministic stress and deformation due to the reactor coolant pump induced pressure pulsations are larger than the random stress and deformation induced by the flow turbulence. The selection of the instruments and measurement locations at Inner barrel assembly top plate in the reactor is essential requirements and very important study process for the vibration and stress measurement program in comprehensive vibration assessment program for the Advanced Power Reactor 1400 reactor vessel internals.